Process of producing cm244 and cm245



PROCESS F PRODUCING Cm AND Cm Winston M. Manning and Martin H. Stndier, Downers Grove, Herbert Diamond, Western Springs, and Paul R. Fields, Chicago, Ill., assignors to the United States of America as represented by the United States Atomic Energy Commission No Drawing. Application September 27, 1957 Serial No. 686,802

3 Claims. (Cl. 233-145) This invention deals with a method of making curium isotopes having mass numbers 244 and 245 and mixtures thereof. It includes a process of producing energy and a method of preparing a material for this purpose.

Curium isotopes of atomic masses between 240 and 242 have been produced heretofore, and their production forms the subject matter of copending application Serial No. 75,064, filed by Glenn T. Seaborg on February 7, 1949.

It is an object of this invention to provide curium isotopes suitable as fuel material for nuclear reactors.

It is another object of this invention to provide a process for the production of curium fertile material and curium thermal-neutron-fissionable material.

It is also an object of this invention to provide a novel method of producing energy.

These objects of the invention are best accomplished by exposing Pu to a high neutron flux, for instance, to a flux of at least 10 neutrons/cm. sec. for a time to obtain an integrated flux in the order of 10 neutrons/cmfl. The irradiated mass is then dissolved and converted to a hydrochloric acid solution of the elements, the remaining plutonium is oxidized to at least its tetravalent state, and the solution is passed through an anion exchange resin whereby the plutonium is absorbed while the fission products and transplutonium values pass through the resin in the effluent. This efiiuent is converted to a dilute hydrochloric acid solution and flowed through a cation exchange resin whereby the transplutonium values are absorbed while some of the fission products remain in the aqueous solution leaving the resin. The cation exchange resin is then washed with dilute hydrochloric acid whereby some fission products, but not the curium are removed. The curium is then eluted with concentrated hydrochloric acid, and the eluate thus obtained is further purified by repetition of the above steps. The curium is then precipitated on a lanthanum fluoride carrier. The precipitate is dissolved in dilute hydrochloric acid, and the solution formed thereby is passed through a cation exchange column. An aqueous citrate solution preferably obtained by neutralizing an aqueous percent citric acid with ammonia to a pH value of 3.3 is then passed through the column whereby the transplutonium values are eluted; the eluate is collected fractionally and the transplutonium values, which are eluted in the order of decreasing atomic and mass numbers, are thereby separated.

All anion and cation exchange resins, respectively, are satisfactory for the process of this invention, such as carboxylic and sulfonic type, strongly or weakly basic, resins. For instance, anion exchange resins made according to U. S. Patents No. 2,356,151; No. 2,402,384; No. 2,591,- 573; No. 2,578,937; No. 2,559,529; No. 2,614,099; No. 2,285,750 and No. 2,469,683 and cation exchange resins made according to U. S. Patent No. 2,366,007 were found to be suitable. I

These steps are the essence of the separation process to be used for the neutron-bombarded plutonium, but addiatent tional intermediate procedures or. steps may be used for the purpose of obtaining a higher degree of separation and purification as will be obvious from the example given below. For instance, various precipitation-dissolution cycles may be inserted, such as precipitation of the plutonium-contaminated trausplutonium values on lanthanum fluoride and/or precipitation of these values as the hydroxides.

While Pu can be neutron-bombarded in any form, namely in the form of the element, in the form of a compound, or in the form of an alloy, the use of plutonium as an alloy with aluminum was preferred. The alloy is preferably encased in an aluminum sheath for protective purposes. In this instance, dissolution of the irradiated body is first carried out with an aqueous sodium hydroxide-sodium nitrate solution whereby the bulk of the aluminum is dissolved while hydroxides of the plutonium and of the values formed by the neutron bombardment remain as a precipitate. This dissolving reaction may be initiated by heating; thereafter it is advantageously controlled by cooling the reaction vessel. The hydroxide precipitate is then separated from the aluminate solution, for instance by centrifuging or any other means known to those skilled in the art, and then dissolved in hydrochloric acid.

When Pu or Pu -containing material is bombarded with neutrons, it is consecutively converted by a series of n, 'y-chain reactions to Pu Pu Pu Pu Pu and Fir Pu and Pu fl-decay to Am and Am respectively. The Am when neutron-bombarded, is converted to the Am which in turn fi-decays to Cm Arnericiurn also forms Pu by electron capture.

consecutively to Cm Cm Cm Cm Cm and.

Cm cm andCm probably Cm and Cm were found to B-decay and to form the berkelium isotopes of the corresponding mass numbers. Bk is converted by neutrons to B11 Bk to B11 and Bk to 31: All four berkelium isotopes ,B-decay and thereby form the corresponding californium isotopes; each californium isotope, when bombarded by neutrons, is converted to the next higher californium istotope. By this, californium isotopes from mass number 250 through 254 are formed. Among these the isotopes Cf and Cf are spontaneously fissionable. Cf ,B-decays and thereby forms element 99 and, more specifically, the isotope of 99 that has the mass number 253. Isotope 99 by neutron bombardment, is converted to 99 and the latter to 99 99 was found to have a half-life of 37 hours and 99 a half-life of 35 days; they ,B-decay and thereby form the element 100, namely, isotopes 100 and 100 which again are converted to 100 by the n, 'y-reaction.

The transplutonic elements which have just been described have been synthesized by neutron-irradiation of Pu and separated from each other. If desired, the berkelium can be isolated from the neutron-irradiated plutonium by dissolving the latter in nitric acid whereby the berkelium is obtained in its tetravalent state and contacting the solution obtained with a dialkyl phosphate, for instance with a heptane solution of di(2-ethyl hexyl) orthophosphoric acid whereby tetravalent actinides and .fission products are extracted away from trivalent actinides and fission products. The organic solution is then scrubbed with a solution of a reducing agent, e. g., with a nitric acid solution containing hydrogen peroxide, whereby berkelium and cesium are reduced to their tri-' valent state and back-extracted away from values of higher valence state. Separation of berkelium from cesium can then be accomplished by selective extraction a of the berkelium with tributyl phosphate or di (Z-ethyl determined. For Cm a half-life'of 19.2:06 years was ascertained, a thermal neutron-capture cross section of 25:10 barns. and a ratio of spontaneous fission-to vat-disintegration of 1:7.25 X 10 Cm has a half-life of about (1.15 05 years and a thermal neutronj-cap m're cross section. of 200i100 barns; it has a thermalneutron-fission; cross section of -'1800 barns. 0m has a half-life of 4000 years and a thermal-neutron capture cross section of 110 barns. V

Cm having a considerably higher thermal-neutronfission' 'cross section than a thermal-neutron-capture cross section and a very long half-life, is highly suitable as a thermal-neutron-fissionable isotope for 'use in a self-sustainirig neutronic reactor for the production of energy or power and/or of fission products having medical and industrial utility. Cm has a low thermal-neutron-cap ture cross section, which, however; is still higher than that of U (2.80). InU containing small quantities of Cm the latter will therefore be converted to the fissionable isotope Crn when neutron-bombarded. For this reason, a neutronic reactor can use a mixture of the twoisotopes (3111 '(fertile isotope) and Cm (fissionable isotope), and separation of'these two isotope's'is not necessary for this use. I least part of the U U or Pu in fuel-elements of neutronic reactors, and the mixture of Cm and Cm can be used in asimilar manner with" natural orenriched uranium, depending upon the'proportion of Cm In thefollowing'an'example is "given for the process of this invention by which the various curium isotopes hav beenproduced in concentrated -form..

EXAMPLE Twentygrams of a sheet of a binary aluminum alloy Cm can thus substitute for at containing 10 percent by weight of Pu was encased in an aluminum sheath. The sheet was bent in the form of a cylinder which had an outside diameter of 1 /2 inches and a height of 2 inches. This cylinder was inserted into a neutronic reactor and exposed there to-neutrons. The

integratedneutron flux"was 8x10" neutrons/cm the flux -3.1:0.5 10 f* 'neutrons/cm. /sec.;- the exposure lasted for about 600 days including various shutdown periods. After this irradiation practically ;all 'of the plutonium except about l0mgxhad been burned out.

The neutron-bombarded cylinder was immersed in 250 ml. of an aqueous solution containing grams of sodium hydroxide and grams of sodium nitrate. Thealuminum was dissolved thereby; after two-fold dilution and heating ofthe solution the plutonium as well as the fission' product and transplutonium values were precipitated as the hydroxides The precipitate was removed bycentrifuging and-then dissolved in'l00 mluof a '10 M aqueous hydrochloric acid to'which-a trace of nitric acid had been added; 'the lattenwa's' for the purpose" of oxidizing the plutonium. to "at leastlits tetravalent state. The anion "exchange resin used throughout: this ex ample was a resin made according to U. S. Patent No. 2,559,529, column 2 line 36 to column 3, line 22; this patenf wa's granted to -Williarn C. Baurnan on July 3,

1951'.; The cationexc'har'ige resin usedfwas one prepared.

in accordance {with Example "1 off-U1 S. Patent No. 366,007 granted to' Gaet'ano 'FZ 'dAlelio' on December- 9 1944:

The hydrochloric acid'solution 'was passed through an anion exchange 'resin'coliimn '1 foothigh and 2 inches in hydrochloric acid-nitric acid mixture which had the same composition as that used'for dissolution of the hydroxide precipitate. 4

The two effluents, that obtained by passing the hydro chloric acid solution through the column and that ob-' tained after washing the resin with the hydrochloric acid nitric acid mixture, were combined and evaporatedto near-dryness. 15 ml. of a 1 M hydrochloric acid was then added, and the solution thus obtained was passed through a cation exchange resinthat was in the form of a column 2 feet high and 1 /2 inches in diameter. The resin was then washed with 300 ml. of a 1 M hydrochloric acid. The efiluent obtained during absorption and that obtained during this last-mentioned washing'step contained some of the fission products.

The transplutonium elements and the trace of residual plutonium were then eluted from the cation exchange resin-by passing ml. of an 11.8 M hydrochloric acid therethrough. The efliuent coming offthe column by. this was collected in20-ml. fractions, and those fractions showing a highalpha activity were then combined. The fractions having low or no alpha-activity mainly contained fis'sionproduets. The alpha-active part of the effluent was then evaporated to a volume of about 3 ml.,

and water'was added thereto in a quantity'to obtain a total volume of 20 ml.

Lanthanum nitrate was then added to the solution in a quantity of 2 mg. and thereafter a 1 M hydrofluoric acid solution whereby" the lanthanum was precipitated as lanthanum fluoride; this precipitate carried practically all of ms mmnium" and transplutonium values. It was separated from the solution by centrifuging and washed with Srnll' of an aqueous solution 1 M in nitric acid and 0.1 M in hydrofluoric acid and thereafter with 4 ml. of 1 Mnitr'ic acid alone. I

The washed precipitate was then dissolvedin 2 ml. of an' 8 M nitric acid that was saturated-with boric acid; after this, water was added to double the volume'of the solution. Diluted ammonium hydroxide was then added to the solution until the latter was neutral and an additional 10% of the quantity needed for neutralizationwas pres ent.- This caused the formation of a hydroxide precipitate; the precipitate was separated from the liquid by centrifuging, washed with water and dissolved in- 10 ml. of an 11.8 M hydrochloric acid that contained a trace of nitric acid.

The solution thus obtained was passed through an anion exchange resin column 3 inches'high and /2 inch wide. Through the'column there were then passe'd25 ml. of an 1l.8 M hydrochloric acid which contained a trace-of nitric-acid whereby the-alpha-activity was washed off the colummj The two effluents, that obtained during absorption-and that obtained-during-wa'shing, were combined and evaporated'tonear-dryness. The evaporation residue was then diluted with 20 ml. of water, and an excessof diluted ammonium hydroxide-was added whereby the transplutonium elements'were precipitated in the form of their'hydroxides. This precipitate was' washed with water.

Thewashedprecipitate was dissolved in 0.350'm1. of 0.5 'Mp'erchloric' acid, and the solution obtained' thereby was passed through a cation exchange resin column 20 cm. highand 4 mm. in diameter. The column container was provided with a jacket through which trichloroethylene vapors were passed to heat the column to 87 C.

Theefiluent coming off the-column was discarded. The

columnwas then washed with an ammonium citrate" M cans acid solutioiito a value of 3.30 with ammonia gas. The effluent was collected 'in fractions of one drop each','and'each drop' was analyzed by alpha-pulse analysisand mass spectrometry.

The total yield of curium fractions was 1.6 x10 alpha counts,'-and the distribution of the 'variou's"-is'ot'opes"' was 5 for Cm 1.36:0.04% for Cm 0.016i0.002% for Cm 0.002% for Cm and 0.002% for Cm The abundance of Cm ions could not be determined on account of interference of the impurity ThO+ of the mass number 248.

As has been mentioned before, material of this composition is suitable as fuel in a nuclear reactor to replace natural uranium. Material containing an enriched proportion (obtained by known processes for separating neutron-bombarded Pu of the thermal-neutron-fissionable Cm is superior to both, U and plutonium, as a nuclear fuel.

The important characteristics for this purpose are the thermal-neutron-fission cross section and the thermalneutron-capture cross section of the isotope or isotopic mixture.

If the ratio of the thermal-neutron-fission cross section to the thermal-neutron-capture cross section is large enough, the number, of neutrons produced and released by fission per each thermal neutron absorbed in the fuel material is greater than unity and a chain reaction is possible.

The calculation of 1;, when the characteristics mentioned above are known, is described in Glasstone and Edlund, the Elements of Nuclear Reactor Theory, Van Nostrand, 1952, chapter 4, and especially on page 83, in equation 4.57.1. By this equation it is possible to calculate the value of 1; for any mixture of curium isotopes.

In the following calculation the isotopes Cm Cm and higher isotopes have been neglected because they have comparatively small macroscopic absorption cross sections, and also because they are present in a small proportion only in the isotopic mixture obtained by the irradiation process described. The number, 1 of neutrons produced per curium fission, extrapolating from present known data of other heavy elements, should be at least 3. According to the calculation referred to above 21(245) 2 too) Z. M244) N 245 X }(245) X v N zrsl mrs) 4245)] 'l' N 244 mm) mm) X I N244 U-245) new new wherein 22;, E 2,, are'the macroscopic cross sections for fission, capture and absorption reactions, respectively, (see Glasstone, Section 3.42), while q, w and a, are

the respective microscopic cross sections; the latter are Therefore For natural uranium the corresponding value is 1.32 (see for example Murray, Nuclear Reactor Principles, Prentice-Hall, 1954, page 106).

Thus the curium isotope mixture obtained in the example behaves, when used as a nuclear fuel, in about the same way as natural uranium. If the Cm isotopes 244 and 245 are still further separated, a higher value for j 6 i the ratio of neutrons formed by fission: neutrons absorbed, is obtained.

An example of a reactor is described in U. S. Patent No. 2,708,656 granted to Fermi and Szilard on May 17, 1955. See the Illustrative Gas Cooled Neutronic Reactor, columns 3742, which deals with the Clinton pile or Oak Ridge Graphite Reactor. This reactor has also been described in less detail in unclassified publications such as Nucleonics, February 1952. A suitable fuel element for such a reactor using the invention contains curium with an isotopic composition of at least 1.25 percent by weight of Cm and at least 95.5 percent by weight of Cm It is preferably used in the form of metal cylinders 1.1 inches in diameter and 4 inches long which are covered with an about 20-mil thick aluminum jacket so that good heat-conductivity with the curium is obtained.

The curium mixture obtained by irradiation and separation, as described, can be readily enriched in the fissionable isotope Cm by electromagnetic separation in the calutron known to those skilled in the art. Because of the high fission-to-capture-cross-section ratio the efliciency of the isotopic mixture as a fuel rises much more rapidly with enrichment in Cm than is the case for a uranium isotopic mixture with increasing enrichment in U or U For example, a mixture of curium isotopes containing 50% Cm and 50% Cm gives a value of at least 2.67 for 7,; a thermal so-called breeder reactor, in which more fissionable material is produced than consumed, can be readily constructed of such fuel.

Such an enriched curium material can also be used to increase the K factor (neutron reproduction factor) in a portion of the lattice of a heterogeneous reactor, in order to compensate for the decreased K factor in other portions of the reactor because of fuel depletion, build-up of poisons, etc., and thereby to maintain the over-all K factor at a level greater than unity. Such enriched curium material can also be used in annular zones near the periphery of the reactor core to build up or flatten the neutron flux distribution and increase the over-all power capability. Other similar uses for such a material will be readily suggested to a person skilled in the art. This increase of the K factor in a portion of a reactor is also described in the above-cited Fermi et a1. patent in columns 44-48.

Another embodiment of the present invention is concerned with the production of californium isotopes which have a high rate of spontaneous fission.

An object of this embodiment is the production of a particularly desirable type of neutron source.

It has been found in accord with this invention that certain californium isotopes are produced by the irradiation of Pu with neutrons of thermal energy in a highflux level for a suitable period of time. The flux level must be at least about 10 neutrons/cmF/sec.

Not only must the flux level or neutron irradiation be high, but the plutonium must be maintained in the flux for a considerable time to achieve a high integrated flux. While microscopic amounts of the californium isotopes of interest can be produced by an integrated flux as low as 4 10 neut'rons/cmfi, in order to obtain useful amounts of Cf and CF the Pu must be exposed to an integrated neutron flux of at least 10 to 10 neutrons/ cmfl.

By the foregoing method californium isotopes of mass numbers 249, 250, 251, 252, 253, and 254 are produced. The present invention however is particularly concerned with the production of the californium isotopes of mass numbers 252 and 254. While all of the isotopes of californium, namely 249, 250, 251, 252, 253, and 254 which are produced by the method of the present invention do spontaneously fission, the half-lives for spontaneous fission of the californium isotopes, except for 252 and 254, are relatively long, so that their usefulness as neutron pro ducers are limited. Cf and CF have short spontaneous fission half-lives of 60 years and 55 days, respectively,

7 so that 1these. isotopes are ..especially suitable .for use .in neutron sources. Californium 252 in. particular can the produced i by the 1 above methods in relatively large ;quantities.

.All of the .californium isotopes produced by :the in radiation of Hu can be separated together from the other 1, transuranic isotopes by the methods previously described in the presentzapplication. The-californium isotopes :of particularinterest'may then be separated'from the other v,californium isotopes by mass-speetrographic 5 the very large neutron sources, which can suitably be a used to replace many of the functions of the conventional nuclear reactor.

Anotherneutron source-utilizes the alpha particles of californium by immersing it-in heavywater with which theyreact bythe :,41 reaction. If enough californium is used, theflux .can be built up to 1x10 neutrons/ cmP/sec. A mixture of curium and beryllium can also be used for this purpose.

Californium is furthermore useful as the radioactive source in atomic batteries like that described in Chemical and Engineering News, 32,592 (1954). A base member of a germanium-antimony or a silicon-antimony alloy with a surface layer of-californium has been found satisfactory for this purpose.

Californium alsocan be added to the fuel material of a neutronic reactor totincrease its IQ-factor, for instance, to a reactor that has a K-factor of about 21.001. Theuse ofcalifornium lin a homogeneous reactor-can make it self-controllable; the expansion of .the waterzis used to shut the reactor off.

Californium is alsouseful as a point source of neutrons in exponential measurements of reactivity of reactorlattices. The mathematics'is much simplified by the fact that californium can be used in a small amount approximating apoint. sThe-californium in the form .of-asphere of viz. 1 mm. diameter encased in zirconium is used assthe point source.

The nuclear propertiesiof-the various californium isotopes produced .by the;present method are shown in the following table.

.Table Isotope .Radia- Halt-life Energy Spontaneous tion 'fission half-lite .oz 550:1:150 yrs. 5.4 :1:0;1 10 years. a 9.45:2.3 yrs.-- "6. 03 0.01 years. a 2.l:t;0.4 yrs". 6.2 01 :1:12 years. 8 183:3 days 55:1:10 days.

been detected.

lttwillbe understood that this invention is not to-be This .application is a continuation-in-part of our 00- pending application Serial 'No. 494,580, filedon Mareh 15, ;1955, nowfabandoned.

'What is 'claimedtis:

1. .A; process ofgprodncing Cm and Cm comprising bombarding vPu -containing material with neutrons of ;thermal.energy; and aflux ofat'least 10 neutrons/cm.

sec. for a time sufficient to obtain an integrated flux of at least 10 neutrons/cm. dissolving the bombarded material in hydrochloric acid,.oxidizing the,plutonium .to 'at .least its .tetravalent' state, flowing :the solution through an anion exchangeresin whereby said plutonium values are adsorbed while the fission products andtransplutonium elements formed remain inithe aqueous solution leavingsaid resin-as eflluent, contactingisaid efiluent .with a cation exchangelresin-whereby said values of'the transplutonium elementsare adsorbed While said 'fission product .values remain in the aqueous solution leaving the resin, contacting the cation exchange resin with an aqueous citrate solution whereby'said transplutonium elements areeluted in the order-of decreasing atomic weight, and collecting said transplutonium-containing citrate eluates fractionally.

2. A process ofproducing Cm and Cm comprising encasinga Pu -containing aluminumv alloy inaluminum, bombarding said alloy with neutrons of thermal energy anda fluxof atleast 10 neutrons/cm. /sec. for atime sufficient to obtainan integrated flux of at least 10 neutrons/cmP, immersing the alloy in y a sodiumhydroxide- .and sodium nitrate-containing aqueous solution whereby the aluminum is .dissolved, adding sodium hydroxide to said solution whereby the plutonium, 'fission products and transplutoniumelements are precipitatedas the hydroxides, separating theprecipitate-fromthe solution, dissolving.theprecipitatein hydrochloric acid, oxidizing the plutonium to at least its tetravalent state, flowing the solution through an anion exchange resin whereby said plutonium values are adsorbed whilethefission products and transplutonium elements formed remain in the aqueous solution leaving said resin as efiiuent, contacting said eflluent with a:cation exchange resin whereby said values of the transplutonium elements are adsorbed while said fission product values remain in the-aqueous solution leaving the resin, contacting the cation exchange resin with an aquous citrate solution whereby said transplutonium elements are eluted in the order of decreasing atomic weight, and collecting said transplutonium-containing citrate eluates fractionally.

3. The process of claim 2 in which the alloy contains about 10% Pu and Al.

References Cited inthe file ofthis'patent .UNITED STATES :PATENTS Seaborg 'et al. Dec. 28, 1954 Street June 21,1955

OTHER REFERENCES Proceedings of the International Conference on the Peaceful Uses of Atomic Energy, United Nations,:N. Y.', vol. 7, Library date Apr. 12, 1956; pages 113-119. 

1. A PROCESS OF PRODUCING CM244 AND CM245 COMPRISING BOMBARDING PU239-CONTAINING MATERIAL WITH NEUTRONS OF THERMAL ENERGY AND A FLUX OF AT LEAST 1014 NUTRONS/CM.2/ SEC. FOR A TIME SUFFICIENT TO OBTAIN AN INTEGRATED FLUX OF AT LEAST 1021 NEUTRONS/CM.2, DISSOLVING THE BOMBARDED MATERIAL IN HYDROCHLORIC ACID, OXIDIZING THE PLUTONIUM TO AT LEAST ITS TETRAVALENT STATE, FLOWING THE SOLUTION THROUGH AN ANION EXCHANGE RESIN WHEREBY SAID PLUTONIUM VALUES ARE ADSORBED WHILE THE FISSION PRODUCTS AND TRANSPLUTONIUM ELEMENTS FORMED REMAIN IN THE AQUEOUS SOLUTION LEAVING SAID RESIN AS EFFLUENT, CONTACTING SAID EFFLUENT WITH A CATION EXCHANGE RESIN WHEREBY SAID VALUES OF THE TRANSPLUTONIUM ELEMENTS ARE ADSORBED WHILE SAID FISSION PRODUCT VALUES REMAIN IN THE AQUEOUS SOLUTION LEAVING THE RESIN, CONTACTING THE CATION EXCHANGE RESIN WITH AN AQUEOUS CITRATE SOLUTION WHEREBY SAID TRANSPLUTONIUM ELEMENTS ARE ELUTED IN THE ORDER OF DECREASING ATOMIC WEIGHT, AND COLLECTING SAID TRANSPLUTONIUM-CONTAINING CITRATE ELUATES FRACTIONALLY. 